1. Field of the Invention
The present invention relates to a method for processing used nuclear fuels, which has, among other advantages, that of avoiding on the site where these fuels are processed, storage of purified plutonium (i.e. totally decontaminated from fission products), even mixed with uranium or uranium and neptunium.
This method notably finds application in the processing of uranium oxide fuels and of mixed uranium and plutonium oxide fuels.
2. Description of the Related Art
Factories for processing used nuclear fuels today use the PUREX (for Plutonium Uranium Refining by EXtraction) method for recovering uranium and plutonium present in these fuels.
This is obtained by applying several purification cycles by liquid-liquid extraction. The extractant used is tri-n-butyl phosphate which has particularly affinity for uranium and plutonium.
The PUREX method, as it is applied in the factory UP2-800 of La Hague in France, schematically comprises three cycles, i.e.:                a first cycle which aims at decontaminating together uranium and plutonium from two actinides(III), americium and curium, as well as from the major portion of fission products, and at achieving a partition of the uranium and of the plutonium into two flows; and        two complementary cycles, designated as <<second uranium cycle>> and <<second plutonium cycle>> respectively and which aim at separately purifying the uranium and plutonium after their partition.        
Recently an important development of the PUREX method, designated as COEX™ (for COEXtraction) method, was proposed in the PCT international application published under the number WO 2007/135178 [1].
Indeed, this development while ensuring uranium and plutonium recovery and purification comparable with those obtained in the PUREX method, gives the possibility of considerably reducing the risks of misappropriation of plutonium for military purposes.
It also allows production of a flow containing a mixture of plutonium, uranium and possibly neptunium totally decontaminated from fission products and the use of this flow for supplying a workshop, a so-called <<co-conversion>> workshop, the function of which is to prepare by oxalic precipitation, a mixed oxide (U,Pu)O2 or (U,Pu,Np)O2 which may be directly used for making nuclear fuels of the MOX (Mixed OXide fuel) type.
To do this, the COEX™ method after operations for decontaminating the uranium and the plutonium similar to the ones applied in the first cycle of the PUREX method, intends to achieve partition of uranium and of plutonium so as to obtain a first flow containing plutonium, uranium and, possibly neptunium, and a second flow containing uranium and possibly neptunium but not containing any plutonium.
It also intends to maintain, in all the operations located downstream from this partition, the plutonium in the presence of uranium and possibly of neptunium until a mixed oxide (U,Pu)O2 or (U,Pu,Np)O2 is obtained.
Thus, in the COEX™ method, the <<second plutonium cycle>> of the PUREX method is replaced with a cycle which aims at purifying the plutonium and uranium and if necessary the neptunium present in the first flow stemming from the partition, from fission products which may still be present in this flow.
In order to ensure decoupling between the operation of the workshops responsible for the processing of used nuclear fuels and that of workshops responsible for making MOX nuclear fuels and to thereby avoid that the production of MOX nuclear fuels is subject to processing of the used nuclear fuels, the COEX™ method comprises a storage step.
This storage is located either between the cycle aiming at purifying the plutonium and uranium and if necessary the neptunium present in the first flow stemming from the partition and the co-conversion, i.e. just after co-conversion.
In the first case, this is a mixture of plutonium, uranium and possibly neptunium purified in an aqueous solution which is stored, while in the second case, this is a mixture of plutonium, uranium and possibly neptunium purified in a solid form which is stored.
With the perspective of making misappropriation of plutonium for military purposes even more difficult during the processing of used nuclear fuels, the Inventors set the goal of providing a method which, while having the same advantages of the COEX™ method, i.e.:                never leaving plutonium alone,        producing a flow containing a mixture of plutonium, uranium and possibly neptunium free of fission products, capable of being converted into a mixed oxide (U,Pu)O2 or (U,Pu,Np)O2 by oxalic co-precipitation, and        ensuring functional decoupling between the processing of used nuclear fuels and the making of new nuclear fuels,further has the advantage of avoiding that purified plutonium, even mixed with uranium or with uranium and neptunium, be stored on the site where the used nuclear fuels are processed.        
They also set the goal that accumulation of these advantages leads to a method, the industrial exploitation of which is not more any complex than that of the COEX™ method and even gives the possibility of obtaining, if possible, a simpler method to be applied industrially than the latter.